ML15126A411

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Initial Exam 2015-301 Draft Administrative Documents
ML15126A411
Person / Time
Site: Sequoyah  Tennessee Valley Authority icon.png
Issue date: 05/06/2015
From:
NRC/RGN-II
To:
Tennessee Valley Authority
References
Download: ML15126A411 (38)
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Text

ES-401 PWR Examination Outline Form ES-401-2 I Facilitv: SEQUOYAH Date of Exam: MARCH 2015 RO KIA Cateoorv Points SRO-Onlv Points Tier Group K K K K A A A A2 1 2 3 K

4 5 K

6 1 2 A

3 4 ..

G Total G* Total

1. 1 3 3 3 3 3 3 18 3 3 6 Emergency&

Abnormal 2 1 1 2 NIA 2 2 NIA 1 9 2 2 4 Plant Evolutions Tier Totals 4 4 5 5 5 4 27 5 5 10 1 3 3 3 3 3 2 2 2 2 2 3 28 3 2 5 2:

Plant 2 0 1 1 1 1 1 1 1 1 1 1 10 0 2 1 3 Systems Tier Totals 3 4 4 4 4 3 3 3 3 3 4 38 5 3 8

3. Generic Knowledge and Abilities 1 2 3 4 10 1 2 3 4 7 Categories 3 3 2 2 1 2 2 2 Note: 1. Ensure that at least two topics from every applicable KIA category are sampled within each tier of the RO and SRO-only outlines (i.e., except for one category in Tier 3 of the SRO-only outline, the "Tier Totals" in each KIA category shall not be less than two).
2. The point total for each group and tier in the proposed outline must match that specified in the table.

The final point total for each group and tier may deviate by +/-1 from that specified in the table based on NRC revisions.

The final RO exam must total 75 points and the SRO-only exam must total 25 points.

3. Systems/evolutions within each group are identified on the associated outline; systems or evolutions that do not apply at the facility should be deleted and justified; operationally important, site-specific systems/evolutions that are not included on the outline should be added. Refer to Section D.1.b of ES-401 for guidance regarding the elimination of inappropriate KIA statements.
4. Select topics from as many systems and evolutions as possible; sample every system or evolution in the group before selecting a second topic for any system or evolution.
5. Absent a plant-specific priority, only those KIAs having an importance rating (iR) of 2.5 or higher shall be selected.

Use the RO and SRO ratings for the RO and SRO-only portions, respectively.

6. Select SRO topics for Tiers 1 and 2 from the shaded systems and KIA categories.

7.* The generic (G) KIAs in Tiers 1 and 2 shall be selected from Section 2 of the KIA Catalog, but the topics must be relevant to the applicable evolution or system. Refer to Section D.1.b of ES-401 for the applicable KIAs.

8. On the following pages, enter the KIA numbers, a brief description of each topic, the topics' importance ratings {IRs) for the applicable license level, and the point totals (#)for each system and category. Enter the group and tier totals for each category in the table above; if fuel handling equipment is sampled in other than Category A2 or G* on the SRO-only exam. enter it on the left side of Column A2 for Tier 2, Group 2 (Note #1 does not apply). Use duplicate pages for RO and SRO-only exams.
9. For Tier 3, select topics from Section 2 of the KIA catalog, and enter the KIA numbers, descriptions, IRs, and point totals(#) on Form ES-401-3. Limit SRO selections to KIAs that are linked to 10 CFR 55.43.

ES-401, REV 9 T1G1 PWR EXAMINATION OUTLINE FORM ES-401-2 KA NAME I SAFETY FUNCTION: IA K1 K2 K3 K4 K5 K6 A1 A2 A3 A4 G TOPIC:

RO SRO 008AG2.2.37 Pressurizer Vapor Space Accident/ 3 3.6 4.6 DD D D DD D D DD ~ Ability to determine operability and/or availability of safety related equipment 009EA 1.14 Small Break LOCA / 3 3.4 3.4 o DDDD~DDDD Secondary pressure control 011 EK3.15 Large Break LOCA / 3 4.3 4.4 D D ~ D D DD D DD Criteria for shifting to recirculation mode 022AA2.04 Loss of Rx Coolant Makeup / 2

2. 9 3.8 D D DD D DD --c--=-~-----

~ D D How long PZR level can be maintained within limits 025AG2.1.28 Loss of AHR System I 4

[J D D ~

4.1 4.1 DODOO Knowledge of the purpose and function of major system components and controls.

027AK2.03 Pressurizer Pressure Control System 2.6 2.8 O ~

DODD Controllers and positioners Malfunction I 3 029EK2.06 ATWS / 1 2.9 3.1 D DODD Breakers, relays, and disconnects.

038EK1 .02 Steam Gen. Tube Rupture/ 3 3.2 3.5 ~ DODOO D Leak rate vs. pressure drop 054AA 1.03 Loss of Main Feedwater / 4 3.5 3.7 D ~ O DOD AFW auxiliaries, including oil cooling water supply 055EA 1.02 Station Blackout/ 6 4.3 4.4 o D D Manual ED/G start 056AA2.60 Loss of Off-site Power/ 6 2.7 2.9 open Page 1of2 04/09/2014 10:08 AM

ES-401, REV 9 T1G1 PWR EXAMINATION OUTLINE FORM ES-401-2 KA NAME I SAFETY FUNCTION: IR K1 K2 K3 K4 KS K6 A1 A2 A3 A4 G TOPIC:

RO SRO 057AA2.02 Loss of Vital AC Inst. Bus I 6 3.7 3.8 0 0 0 0 0 0 0 ~ 0 0 0 Core flood tank pressure and level indicators

---m-**--- -*----* *------ *------------------

058AG2.4.49 Loss of DC Power I 6 4.6 4.4 [] 0000 ODO~ Ability to perform without reference to procedures those actions that require immediate operation of system components and controls.

077AK3.01 Generator Voltage and Electric Grid 3.9 4.2 0 0 ~ 0 0 0 0 0 0 0 Reactor and Turbine trip criteria Disturbances I 6 WE04EK1 .3 LOCA Outside Containment I 3 3.5 3.9 ~ 0 0000 ODO Annunciators and conditions indicating signals, and remedial actions associated with the (LOCA Outside Containment).

WE05EK3.4 Inadequate Heat Transfer - Loss of 3.7 3.9 o~o 000000 RO or SRO function within the control room team as Secondary Heat Sink I 4 appropriate to the assigned position, in such a way that procedures are adhered to and the limitations in the facilities license and amendments are not violated.

WE11 EK2.2 Loss of Emergency Coolant Recirc. I 4 3.9 4.3 O ~ 0 OOO

  • ------------------*-*--***---~--------

Facility's heat removal systems, including primary coolant, emergency coolant, the decay heat removal systems and relations between the proper operation of these systems to the operation of the facility.

WE12EK1 .3 Steam Line Rupture - Excessive Heat 3.4 3.7 ~ O 0000 ODO Annunciators and conditions indicating signals, and Transfer I 4 remedial actions associated with the (Uncontrolled Depressurization of all Steam Generators).

Page 2 of 2 04/09/2014 10:08 AM

ES-401, REV 9 T1G2 PWR EXAMINATION OUTLINE FORM ES-401-2 KA NAME I SAFETY FUNCTION: IR K1 K2 K3 K4 K5 K6 A1 A2 A3 A4 G TOPIC:

RO SRO 001AA2.03 Continuous Rod Withdrawal I 1 4.5 4.8 0 0 0 0 0 0 0 ~ 0 0 0 Proper actions to be taken if automatic safety functions have not taken place 028AK3.03 Pressurizer Level Malfunction I 2 3.5 4.1 OO ~ O O [] O0 0 0 False indication of PZR level when PORV or spray valve is open and RCS saturated 051AA1.04 Loss of Condenser Vacuum I 4 2.5 2.5 0 0000~00 Rod position 060AA1.01 Accidental Gaseous Radwaste Rel. I 9 2.8 3 000000~0000 Area radiation monitors

--~~~--~~----~~~~~~~~--

068AK2.01 Control Room Evac. I 8 3.9 4 0~000000 DO Auxiliary shutdown panel layout 076AG2.1.32 High Reactor Coolant Activity I 9 3.8 4.0 0 0 000000~ Ability to explain and apply all system limits and precautions.

0 0 0 0 OOOOOO

~

WE08EK1.2 RCS Overcooling - PTS I 4 3.4 4.0 ~ Normal, abnormal and emergency operating procedures associated with (Natural Circulation Operations).

WE10EA2.2 Natural Circ. With Seam Void/ 4 3.4 3.9 0 00000~0 0 Adherence to appropriate procedures and operation within the limitations in the facility's license and amendments.

WE15EK3.4 Containment Flooding I 5 2.9 3.0 0 0 ~ 0 0 0 ODO RO or SRO function within the control room team as appropriate to the assigned position, in such a way that procedures are adhered to and the limitations in the facilities license and amendments are not violated.

Page 1of1 04/09/2014 10:08 AM

ES-401, REV 9 T2G1 PWR EXAMINATION OUTLINE FORM ES-401-2 KA NAME I SAFETY FUNCTION: IA K1 K2 K3 K4 K5 K6 A1 A2 A3 A4 G TOPIC:

RO SRO 003K5.04 Reactor Coolant Pump 3.2 3.s D DD D ~ D D DDD D Effects of RCP shutdown on secondary parameters, such as steam pressure, steam flow and feed flow 004G2.4.31 Chemical and Volume Control 4.2 4.1 DDDDDDD D~ Knowledge of annunciators alarms, indications or response procedures

*----------- -------------*------------- --*------*-----~*--

0041<2.04 Chemical and Volume Control 2.6 2.7 O ~ O OO O O O OO O BWSTtankheaters 005K4.03 Residual Heat Removal 2.9 3.2 DODD AHR heat exchanger bypass flow control 006K2.01 Emergency Core Cooling 3.6 3.9 D ~ DD DD D D D ECCS pumps 007K5.02 Pressurizer Relief/Quench Tank 3.1 3.4 D DD D ~ D DD D Method of forming a steam bubble in the PZR 008A1.01 Component Cooling Water 2.8 2.9 DD DD DD ~ D DD D CCWflowrate 010K4.01 Pressurizer Pressure 2.7 2.9 n O ~ O O O O OOO valve warm-up 012K3.01 Reactor Protection 3.9 4.0 DD -o 013K1.06 Engineered Safety Features Actuation 4.2 4.4 ~ oo oooor-1 013K2.01 Engineered Safety Features Actuation 3.6 3.8 n ~

[] D

--~------*---------*-****-

ESFAS/safeguards equipment control Page 1of3 04/09/2014 10:08 AM

ES-401, REV 9 T2G1 PWR EXAMINATION OUTLINE FORM ES-401*2 KA NAME I SAFETY FUNCTION: IA K1 K2 K3 K4 K5 K6 A1 A2 A3 A4 G TOPIC:

RO SRO 022A2.03 Containment Cooling 2.6 3.o D D D D D D D ~ D D D Fan motor thermal overload/high-speed operation 022K3.02 Containment Cooling 3.o 3.3 DD~ DODD DD Containment instrumentation readings 025K6.01 Ice Condenser 3.4 3.6 D D D D D ~ D D D D D Upper and lower doors of the ice condenser 026A4.05 Containment Spray 3.5 3.5 D D D D D D~ Containment spray reset switches 039A3.02 Main and Reheat Steam 3.1 3.5 o DDDDD~DD Isolation of the MASS 059A1.07 Main Feedwater 2.5 2.6 DD DD D~DD Feed Pump speed, including normal control speed for ICS 059G2.4.3 Main Feedwater 3.7 3.9 D D D D D D D D D ~ Ability to identify post-accident 061K5.02 Auxiliary/Emergency Feedwater 3.2 3.6 o D~DDDD Decay heat sources and magnitude 061K6.02 Auxiliary/Emergency Feedwater 2.6 2.1 oo D ~ D DD Pumps 062G2.2.39 AC Electrical Distribution 3.9 4.5 n n [] D D D D D D ~ Knowledge of less than one hour technical specification action statements for systems.

063K1.03 DC Electrical Distribution 2.9 3.5 ~ [JD D Battery charger and battery Page 2 of 3 04/09/2014 10:08 AM

ES-401, REV 9 T2G1 PWR EXAMINATION OUTLINE FORM ES-401-2 KA NAME I SAFETY FUNCTION: IR K1 K2 K3 K4 KS K6 A1 A2 A3 A4 G TOPIC:

RO SRO 064K1.02 Emergency Diesel Generator 3.1 3.6 ~ D D D D D D D D D D DIG cooling water system 073K3.01 Process Radiation Monitoring 3.6 4.2 O O ~ O O O O O O O Radioactive releases 076A3.02 Service Water 3.7 3.7 D D D D D D D D ~ D D Emergency heat loads 076K4.03 Service Water 2.9 3.4 D D D ~ D D DODD Automatic opening features associated with SWS isolation valves to CCW heat exchanges 078A4.01 Instrument Air 3.1 3.1 D D DDDDLJ~[.J gauges 103A2.04 Containment 3.5 3.6 DDDD~DDD Containment evacuation (including recognition of the alarm)

Page 3of 3 04/09/2014 10:08 AM

ES-401, REV 9 T2G2 PWR EXAMINATION OUTLINE FORM ES-401-2 KA NAME I SAFETY FUNCTION: IA K1 K2 K3 K4 K5 K6 A1 A2 A3 A4 G TOPIC:

RO SRO 001K2.01 Control Rod Drive 3.5 3.6 D ~ D D D D D D D D D One-line diagram of power supply to MIG sets.

015G2.2.40 Nuclear Instrumentation 3.4 4.7 D D D D D D D D D ~ Ability to apply technical specifications for a system.

016A3.01 Non-nuclear Instrumentation 2.9 2.9 D D D D D D D D ~ D D Automatic selection of NNIS inputs to control systems 028A2.02 Hydrogen Recombiner and Purge 3.5 3.9 D D D D D D D ~ DD LOCA condition and related concern over hydrogen Control

  • -----:c-==~-=--=-~=-==-=

034K6.02 Fuel Handling Equipment 2.6 3.3 D D D D ~ D D D D D Radiation monitoring systems 041K3.02 Steam Dump/Turbine Bypass Control 3.8 3.9 r1 D ~ D D D RCS 071A1.06 Waste Gas Disposal 2.5 2.8 D DDDD~DDDD Ventilation system 072K5.01 Area Radiation Monitoring 2.7 3.0 n n n n ~ D D D D D Radiation theory, including sources, types, units and effects 075K4.01 Circulating Water 2.5 2.8 D D D ~ D Heat sink 086A4.06 Fire Protection 3:2 3.2 D ~ Halon system Page 1of1 04/09/2014 10:08 AM

ES-401, REV 9 T3 PWR EXAMINATION OUTLINE FORM ES-401 *2 KA NAME I SAFETY FUNCTION: IA K1 K2 K3 K4 K5 K6 A1 A2 A3 A4 G TOPIC:

RO SRO G2.1.14 Conduct of operations 3.1 3.1 D D D D D D D D D D ~ Knowledge of criteria or conditions that require plant-wide announcements, such as pump starts, reactor trip, mode changes, etc.

G2.1.34 Conduct of operations 2.1 3.5 DD DDDDDDD~ Knowledge of primary and secondary chemistry limits G2.1.5 Conduct of operations 2.9 3.9 O O O O O O O O O O ~ Ability to locate and use procedures related to shift staffing, such as minimum crew complement, overtime limitations, etc.

G2.2.13 Equipment Control 4.1 4.3 oo DDDDDDD~ Knowledge of tagging and clearance procedures.

G2.2.2 Equipment Control 4.6 4.1 D D D D D D 0 D D D ~ Ability to manipulate the console controls as required to operate the facility between shutdown and designated power levels.

G2.2.41 Equipment 9ontrol 3.5 3.9 D 0 D 0 0 ODDO~ Ability to obtain and interpret station electrical and mechanical drawings G2.3.5 Radiation Control 2.9 2.9 D 0 0 0 DODD ~ Ability to use radiation monitoring systems G2.3.7 Radiation Control 3.5 3.6 0 bll to comply with radiation work permit requirements during normal or abnormal conditions G2.4.39 Emergency Procedures/Plans 3.9 3.8 r-1 11 r1 n n n D D 0 D ~ Knowledge of the RO's responsibilities in emergency plan implementation.

G2.4.4 Emergency Procedures/Plans 4.5 4.7 D D D D D [] [] ~ Ability to recognize abnormal indications for system operating parameters which are entry-level conditions for emergency and abnormal operating procedures.

Page 1of1 04/09/2014 10:08 AM

ES-401, REV 9 SRO T1G1 PWR EXAMINATION OUTLINE FORM ES-401-2 KA NAME I SAFETY FUNCTION: IA K1 K2 K3 K4 KS K6 A1 A2 A3 A4 G TOPIC:

RO SRO 015AA2.02 RCP Malfunctions I 4 2.8 3 DDDDDDD~DDD Abnormalities in RCP air vent flow paths and/or oil cooling system 022AG2.2.4 Loss of Rx Coolant Makeup I 2 3.6 3.6 D DDDDDDD~ (multi-unit) Ability to explain the variations in control board layouts, systems, instrumentation and procedural actions between units at a facility.

029EG2.2.38 ATWS I 1 3.6 4.s o [J DODD~ Knowledge license.

conditions and limitations in the facility 040AG2.4.30 Steam Line Rupture* Excessive Heat 2.7 4.1 O O DD DODD~ Knowledge of events related to system operations/status Transfer I 4 that must be reported to internal orginizations or outside agencies.

057AA2.04 Loss of Vital AC Inst. Bus I 6 3.7 4 D ODDO ~DOD ESF system panel alarm annunciators and channel status indicators

. -~~~~~-

077AA2.01 Generator Voltage and Electric Grid Disturbances I 6 3.s 3.6 DDDDDDD~ D- O Operating point on the generator capability curve Page 1of1 04/09/2014 10:08 AM

ES-401, REV 9 SRO T1G2 PWR EXAMINATION OUTLINE FORM ES-401-2 KA NAME I SAFETY FUNCTION: IR K1 K2 K3 K4 KS K6 A 1 A2 A3 A4 G TOPIC:

RO SRO 001AA2.02 Continuous Rod Withdrawal/ 1 4.2 4.2 D DD DD D D ~ DD D Position of emergency boration valve 033AG2.1.20 Loss of Intermediate Range NI / 7 4.6 4.6 DD DD D ~ Ability to execute procedure steps.

061AA2.06 ARM System Alarms/ 7 s.2 4.1 D DDDDD~DDD Required actions if alarm channel is out of service we14EG2.2.44 Loss of CTMT Integrity/ 5 4.2 4.4 D D D DO O OO O ~ Ability to interpret control room indications to verify the status and operation of a system, and understand how operator actions and directives affect plant and system conditions Page 1 of 1 04/09/2014 10:08 AM

ES-401, REV 9 SRO T2G1 PWR EXAMINATION OUTLINE FORM ES-401 *2 KA NAME I SAFETY FUNCTION: IR K1 K2 K3 K4 K5 K6 A1 A2 A3 A4 G TOPIC:

RO SRO 012G2.4.49 Reactor Protection 4.6 4.4 D D D D D D D D D D ~ Ability to perform without reference to procedures those actions that require immediate operation of system components and controls.

025A2.05 Ice Condenser 2.5 2.1 [J o DDD~DDD Abnormal glycol expansion tank level 039G2.4.50 Main and Reheat Steam 4.2 4.o D D D D D D D D D D ~ Ability to verify system alarm setpoints and operate controls identified in the alarm response manual.

063A2.01 DC Electrical Distribution 2.5 3.2 D D D D D D D ~ D D D Grounds 103A2 03 Containment 3.5 3.8 oo DDD~DDD Phase A and B isolation Page 1of1 04/09/2014 10:08 AM

ES-401, REV 9 SRO T2G2 PWR EXAMINATION OUTLINE FORM ES-401-2 KA NAME I SAFETY FUNCTION: IA K1 K2 K3 K4 K5 K6 A1 A2 A3 A4 G TOPIC:

RO SRO 028G2.1.25 Hydrogen Recombiner and Purge 3.9 4.2 0 0 0 0 0 0 0 0 0 ~ Ability to interpret reference materials such as graphs, Control monographs and tables which contain performance data.

071A2.02 Waste Gas Disposal 3.3 3.6 0 0~000 Use of waste gas release monitors, radiation, gas flow rate and totalizer 079A2.01 Station Air 2.9 3.2 0 0 ~ 0 0 Cross-connection with IAS Page 1of1 04/09/2014 10:08 AM

ES*401, REV 9 SRO T3 PWR EXAMINATION OUTLINE FORM ES-401*2 KA NAME I SAFETY FUNCTION: IA K1 K2 K3 K4 KS K6 A1 A2 A3 A4 G TOPIC:

RO SRO G2.1.35 Conduct of operations 2.2 3.9 D DDDDDDDD~ Knowledge of the fuel handling responsibilities of SRO's G2.2.17 Equipment Control 2.6 3.8 D lJDDDDDDD~ Knowledge of the process for managing maintenance activities during power operations.

G2.2.21 Equipment Control 2.9 4.1 DDDDDDDDDD~ Knowledge of pre- and post-maintenance operability requirements.

G2.3.13 Radiation Control 3.4 3.8 DDDDDD D ~ Knowledge of radiological safety procedures pertaining to licensed operator duties G2.3.6 Radiation Control 2.0 3.8 DDDDDDDDDD~ Ability to aprove release permits G2.4.38 Emergency Procedures/Plans 2.4 4.4 DDDDDDDDDD~ Ability to take actions called for in the facility emergency plan, including supporting or acting as emergency coordinator.

G2.4.5 Emergency Procedures/Plans 3.7 4.3 nnnDDD DD~ Knowledge of the organization of the operating procedures network for normal, abnormal and emergency evolutions.

Page 1of1 04/09/2014 10:08 AM

ES-301 Administrative Topics Outline Form ES-301-1 Facility: Seguoy:ah Nuclear Station 1 & 2 Date of Examination: 3/2/2015 Exam Level: RO 0 sRoD Operating Test No: 2015-301 Administrative Topic (see Type Describe activity to be performed Note) Code*

Perform 1-Pl-OPS-000-020.1 OPERATOR AT THE CONTROLS DUTY STATION CHECKLISTS MODES 1-4 to assess grid status and take required actions.

Conduct of Operations R,N 2.1.7 (4.4/4.7) Ability to evaluate plant performance and make operational judgments based on operating characteristics, reactor behavior, and instrument interpretation.

Determine system lineup from a clearance restoration.

Conduct of Operations R,N 2.1.29 (4.1 /4.1) Knowledge of how to conduct system lineups, such as valves, breakers, switches, etc.

Determine QPTR Equipment Control R,N 2.2.12 (3.0) Knowledge of surveillance procedures.

Radiation Control Not examined Complete a TVA INITIAL NOTIFICATION FOR GENERAL EMERGENCY Emergency Procedures/Plan R,N 2.4.39 (3.9) Knowledge of RO responsibilities in emergency plan implementation.

NOTE: All items (5 total) are required for SROs. RO applicants require only 4 items unless they are retaking only the administrative topics, when all 5 are required.

  • Type Codes & Criteria:(C)ontrol room, (S)imulator, or Class(R)oom (D)irect from bank (S 3 for ROs; : : ; 4 for SROs & RO retakes)

(N)ew or (M)odified from bank (~ 1)

(P)revious 2 exams (S 1; randomly selected)

A.1.a While performing 1-Pl-OPS-000-020.1 OPERATOR AT THE CONTROLS DUTY STATION CHECKLISTS MODES 1-4 and given data that includes a high Switchyard voltage with the Main Generator low, the examinee will refer to Appendix E Guidance for Voltage Schedules and determine that further Main Generator voltage adjustments cannot be made to reduce Switchyard voltage. The examinee will determine the following:

  • Notify SELD (Transmission Operator) within 30 min.
  • Log entry will be made to satisfy NERC requirements.

This JPM is based on Sequoyah plant operating experience.

A.1.b The examinee will determine the restoration of 1B Charging Pump and by using 1-S0-62-1 CHEMICAL AND VOLUME CONTROL SYSTEM.

This JPM is based on Sequoyah plant operating experience.

A.2 The examinee will perform Quadrant Power Tilt Ratio with the plant computer INOPERABLE using O-Sl-NUC-000-133.0. The examinee will gather data, calculate, interpret and determine the acceptance is not met.

A.3 Not examined.

A.4 While acting as the Site Communicator and given data for a plant emergency, the examinee will interpret the data and complete the EPIP-5 GENERAL EMERGENCY Appendix A GENERAL EMERGENCY INITIAL NOTIFICATION FORM within 15 minutes.

ES-301 Administrative Topics Outline Form ES-301-1 Facility: Sequoyah Nuclear Station 1 & 2 Date of Examination: 3/2/2015 Exam Level: RO D SRO 0 Operating Test No: 2015-301 Administrative Topic (see Type Describe activity to be performed Note) Code*

Perform 1-Pl-OPS-000-020.1 OPERATOR AT THE CONTROLS DUTY STATION CHECKLISTS MODES 1-4 to assess grid status and take required actions.

Conduct of Operations R,N 2.1. 7 (4.4/4.7) Ability to evaluate plant performance and make operational judgments based on operating characteristics, reactor behavior, and instrument interpretation.

Review System Lineup From a Clearance Restoration.

Conduct of Operations R,N 2.1.29 (4.1 /4.1) Knowledge of how to conduct system lineups, such as valves, breakers, switches, etc.

SRO Determine Mode Change Requirements with INOP equipment Equipment Control R,N 2.2.35 (4.5) Ability to determine Technical Specification Mode of Operation.

Select and provide approval for workers to exceed their Administrative Dose Limit in order to make an emergent entry into the RCA.

2.3.13 (3.8) Knowledge of radiological safety Radiation Control R,M procedures pertaining to licensed operator duties, such as response to radiation monitor alarms, containment entry requirements, fuel handling responsibilities, access to locked high-radiation areas, aligning filters, etc.

Classify the event and determine Protective Action Recommendations.

Emergency Procedures/Plan R,M 2.4.41 (4.6) Knowledge of the emergency action level thresholds and classifications.

NOTE: All items (5 total) are required for SROs. RO applicants require only 4 items unless they are retaking only the administrative topics, when all 5 are required.

  • Type Codes & Criteria:(C)ontrol room, (S)imulator, or Class(R)oom (D)irect from bank (:S 3 for ROs; :S 4 for SROs & RO retakes)

(N)ew or (M)odified from bank (~ 1)

(P)revious 2 exams (:S 1; randomly selected)

A.1.a While performing 1-Pl-OPS-000-020.1 OPERATOR AT THE CONTROLS DUTY STATION CHECKLISTS MODES 1-4 and given data that includes a high Switchyard voltage with the Main Generator low, the examinee will refer to Appendix E Guidance for Voltage Schedules and determine that further Main Generator voltage adjustments cannot be made to reduce Switchyard voltage. The examinee will determine the following:

  • Notify SELD (Transmission Operator) within 30 min.
  • Log entry will be made to satisfy NERC requirements.

This JPM is based on Sequoyah plant operating experience.

A.1.b The examinee will determine the restoration of 1B Charging Pump and by using 1-S0-62-1 CHEMICAL AND VOLUME CONTROL SYSTEM.

This JPM is based on Sequoyah plant operating experience.

A.2 Given Unit 1 is in Mode 4 with the following equipment INOPERABLE:

  • RCS PZR Level Instrument Loop (1-L-68-320) for emergent repair.
  • Power Range NI Channel I (N-41) removed from service for maintenance.

and preparations are being made to perform a change to MODE 3, the examinee will review NPG-SPP-09.11.2 Risk Assessment Methods for Technical Specifications and Unit 1 Technical Specifications and assess given the rules of usage that the mode change to MODE 3 can proceed.

A.3 Given seven different workers during a lifesaving emergency, the Examinee will choose two workers of the five for the authorization to exceed the TVA Administrative Dose Levels, determine authorization from the Site Emergency Director is required and the maximum TEDE exposure limit for the conditions given is 25 Rem.

A.4 While acting as the Site Emergency Director and given data for a plant emergency, the examinee will use EPIP-1 EMERGENCY PLAN CLASSIFICATION MATRIX to interpret the data within 15 minutes and determine the correct Emergency Classification of General Emergency.

The examinee will then use EPIP-5 GENERAL EMERGENCY to determine Protective Action Recommendation (Recommendation 2).

ES-301 Control Room/In-Plant Systems Outline Form ES-301-2 Facility: Sequoyah Nuclear Station 1 & 2 Date of Examination: 3/2/2015 Exam Level: RO -.J SRO-I D SRO-U D Operating Test No: 2015-301 Control Room Systems (8 for RO); (7 for SRO-I); (2 or 3 for SRO-U, including 1 ESF)

System I JPM Title Type Code* Safety Function

a. Continuous Rod Withdrawal While Withdrawing Shutdown Bank N, A, S 1 Rods During a Startup. APE 001 AA 1.05 (4.3/4.2)
b. Depressurize the RCS to Enable Safety Injection Pump Flow N, EN,L,S 2 During a Loss of Charging. APE 022AA1.03 (3.2/3.2)
c. Isolate the Ruptured Steam Generator with MSIV Failure to D,A,L,S 3 Close. EPE 038 EA1.32 (4.6/4.7)
d. Start a Reactor Coolant Pump APE 015/017 AA2.09 (3.4/3.5) D, A, L, S 4P
e. Initiate AFW Flow to Steam Generators While in ECA-2.1. EPE N,S 4S E12 EA 2.2 (3.4/3.9)
f. Respond to High Containment Pressure, Place RHR Spray in D, A, EN, L, 5

Service. EPE E 14 EA 1.1 (3. 7/3. 7) s

g. Respond to a Loss of 1B Shutdown Board. 062 A2.04 (3.1/3.4) N,A,S 6
h. Respond to ERCW Pump Trip and ERCW Rupture. APE AO?

M,A,S 8 AA 1.3 (3.3/3.5)

In-Plant Systems (3 for RO); (3 for SRO-I); (3 or 2 for SRO-U)

i. Close MSIV's Locally. APE 040AA1.04 (4.3/4.3) N,E,L 4S
j. Take Local Control of the Turbine Driven AFW Level Control N,R,E,L 8 valve LCV-3-175. APE 068 AA1.03 (4.1/4.3)
k. Perform Radiation Monitor O-RE-90-122 Flushing. 068 A4.04 D,R 9 (3.8/3.7)

@ All RO and SRO-I control room (and in-plant) systems must be different and serve different safety functions; all 5 SRO-U systems must serve different safety functions; in-plant systems and functions may overlap those tested in the control room.

  • Type Codes Criteria for RO I SRO-I I SRO-U (A)lternate path 4-6 I 4-6 I 2-3 (C)ontrol room (D)irect from bank S9/S8/S4 (E)mergency or abnormal in-plant <::1/<::1/<::1 (EN)gineered safety feature - I - I <::1 (control room system)

(L)ow-Power I Shutdown <::1/<::1/<::1 (N)ew or (M)odified from bank including 1(A) <::2/<::2/<::1 (P)revious 2 exams  :::; 3 I:::; 3 I :::; 2 (randomly selected)

(R)CA <::1/<::1/<::1 (S)imulator

a. Shutdown Bank Withdrawal with Continuous Rod Motion The examinee will assume the shift in MODE 3 with prerequisites complete to withdraw Shutdown Bank "A" using O-S0-85-1 CONTROL ROD DRIVE SYSTEM to initialize the Rod Control Startup Step Counter Reset, close the Reactor Trip breakers and momentarily withdraw Shutdown Bank A.

When Shutdown Bank "A" is withdrawn a continuous rod motion condition occurs. The examinee will trip the Reactor using the Immediate Operator Actions of AOP-C.01, ROD CONTROL SYSTEM MALFUNCTIONS.

b. Depressurize the RCS to Enable Safety Injection Pump Flow During a Loss of Charging The examinee will assume the shift with the Reactor tripped and ES-0.1, Reactor Trip Response complete following a loss of all Charging pumps. The examinee will use AOP-M.09, LOSS OF CHARGING starting at step 27 to block Safety Injection and commence RCS Pressure reduction using a PORV to enable Safety Injection flow as demonstrated by Pressurizer level increasing.
c. Isolate the Steam Generator Tube Rupture (With MSIV Failure to Close)

The examinee will assume the shift with the Reactor tripped following a Steam Generator Tube Rupture. The examinee will use E-3, STEAM GENERATOR TUBE RUPTURE to attempt to isolate a ruptured steam generator using the ruptured steam generator MSIV. The ruptured steam generator MSIV will fail to close requiring the examinee will to isolate the steam paths of the ruptured Steam Generator by closing all intact Steam Generator MSIV's and the alternate flowpath isolation valves.

d. Start a Reactor Coolant Pump The examinee will assume the shift in MODE 3 with prerequisites complete to start a Reactor Coolant pump. The examinee will use 1-S0-68-2 REACTOR COOLANT PUMPS to start the Reactor Coolant Pump (RCP). Shortly after starting the pump, a high motor winding condition develops. The examinee will use Alarm Response 1-AR-M5-B and will transition to AOP-R.04 REACTOR COOLANT PUMP MALFUNCTIONS to manually trip the RCP.
e. Initiate AFW Flow to Steam Generators While in ECA-2.1 The examinee will assume the shift during a common fault that results in all Steam Generators depressurizing. The examinee will use ECA-2.1, UNCONTROLLED DEPRESSURIZATION OF ALL STEAM GENERATORS to close the TD AFW pump steam supply valve and to reduce AFW flow to each Steam generator to 50 :!: 1Ogpm.
f. Respond To High Containment Pressure Condition By Placing RHR Spray In Service.

The examinee will assume the shift during a LOCA and a High Containment Pressure condition. The examinee will use FR-Z.1, High Containment Pressure, starting at step 13 to establish RHR Spray flow to reduce Containment pressure. During the alignment, the B Train RHR Spray flow isolation valve FCV-72-41 fails to operate. The examinee uses the alternate path to re-establish B Train RHR to the injection mode and establishes A Train RHR to the Containment Spray mode.

g. Respond to a Loss of Shutdown Board The examinee will assume the shift in MODE 1 when a loss of power to the 1B 6.9kV Shutdown Board occurs. The examinee will use AOP-P.05 LOSS OF UNIT 1 SHUTDOWN BOARDS to evaluate RCP seal cooling and start 1A Charging pump. The examinee will evaluate the 1B Diesel generator and determine the Diesel requires an EMERGENCY STOP due to a failure of the cooling water valves to open. Finally, the examinee will stop all AFW flow to the SG from the TD AFW pump.
h. Respond to ERCW Pump Trip and Isolate the ERCW Leak The examinee will assume the shift in MODE 1 when a trip of the Q-A Pump occurs, the examinee will manually start the J-A ERCW pump using AOP-M.01, LOSS OF ESSENTIAL RAW COOLING WATER. After the pump is started, the examinee will evaluate ERCW flow and determine a rupture has occurred in "A" Train ERCW. The examinee will subsequently manually place all "A" Train ERCW pumps in PTL and open all alternate ERCW supply to DG valves.
i. Close MSIV's Locally The examinee will assume the shift with the Reactor tripped following a fault on the Unit 1 #1 Steam Generator with a failure of the MSIV's to close. The examinee will use EA-1-1, CLOSING MSIVs LOCALLY to simulate placing the Auxiliary Control Room transfer switches to AUX and subsequently simulate removing the control power fuses from 125V Vital Battery Boards I and 11 for the failed open MSIV.
j. Take Local Control of the Turbine Driven AFW Level Control valve LCV-3-175 The examinee will assume the shift following a Control Room evacuation with the Control Room staff unable to control Auxiliary Feed to the Unit 2 #4 Steam Generator. The examinee will use AOP-C.04 SHUTDOWN FROM AUXILIARY CONTROL ROOM, APPENDIX W.2 CONTROL OF UNIT 2 TURBINE DRIVEN AFW FLOW FROM OUTSIDE MCR to simulate aligning the Station Blackout air bottles to the failed open LCV-3-175, Unit 2 #4 Steam Generator level control valve and control Auxiliary Feed flow to the Unit 2 #4 Steam Generator. This JPM reflects a new plant modification.
k. Radiation Monitor O-RE-90-122 Flushing After Hi Radiation Signal Isolation.

The examinee will assume the shift following a high radiation isolation during Cask Decon Collector Tank release. The examinee will use O-S0-77-1 WASTE DISPOSAL SYSTEM (LIQUID), section 8.2 Radiation Monitor O-RE-90-122 Flushing After Hi Radiation Signal Isolation of Release to simulate flushing the Liquid Effluent Radiation Monitor in an attempt to clear the high radiation signal. The examinee will then simulate isolating the Cask Decon Collector Tank and flushing the Cask Decon Collector Tank discharge to the Floor Drain Collector tank. The examinee will then simulate re-aligning the Cask Decon Collector Tank for liquid release.

ES-301 Control Room/In-Plant Systems Outline Form ES-301-2 Facility: Sequoyah Nuclear Station 1 & 2 Date of Examination: 3/2/2015 Exam Level: RO D SRO-I "./ SRO-U D Operating Test No: 2015-301 Control Room Systems (8 for RO); (7 for SRO-I); (2 or 3 for SRO-U, including 1 ESF)

System I JPM Title Type Code* Safety Function

a. Continuous Rod Withdrawal While Withdrawing Shutdown Bank N, A, S 1 Rods During a Startup. APE 001 AA1.05 (4.3/4.2)
b. Depressurize the RCS to Enable Safety Injection Pump Flow N, EN, L, S 2 During a Loss of Charging. APE 022AA1.03 (3.2/3.2)
c. Isolate the Ruptured Steam Generator with MSIV Failure to D, A, L, S 3 Close. EPE 038 EA1.32 (4.6/4.7)
d. Start a Reactor Coolant Pump APE 015/017 AA2.09 (3.4/3.5) D, A, L, S 4P
e. Not examined N/A N/A
f. Respond to High Containment Pressure, Place RHR Spray in D, A, EN, L, 5

Service. EPE E 14 EA 1.1 (3. 7/3. 7) s

g. Respond to a Loss of 1 B Shutdown Board. 062 A2.04 (3.1/3.4) N, A, S 6
h. Respond to ERCW Pump Trip and ERCW Rupture. APE AO?

M, A, S 8 AA 1.3 (3.3/3.5)

In-Plant Systems (3 for RO); (3 for SRO-I); (3 or 2 for SRO-U)

i. Close MSIV's Locally. APE 040 AA1.04 (4.3/4.3) N, E,L 4S
j. Take Local Control of the Turbine Driven AFW Level Control N, R, E, L 8 valve LCV-3-175. APE 068AA1.03 (4.1/4.3)
k. Perform Radiation Monitor O-RE-90-122 Flushing. 068 A4.04 D,R 9 (3.8/3.7)

@ All RO and SRO-I control room (and in-plant) systems must be different and serve different safety functions; all 5 SRO-U systems must serve different safety functions; in-plant systems and functions may overlap those tested in the control room.

  • Type Codes Criteria for RO I SRO-I I SRO-U (A)lternate path 4-6 I 4-6 I 2-3 (C)ontrol room (D)irect from bank s;g/s;8/s;4 (E)mergency or abnormal in-plant ~1/~1/~1 (EN)gineered safety feature - I - I ~1 (control room system)

(L)ow-Power I Shutdown ~1/~1/~1 (N)ew or (M)odified from bank including 1(A) ~2/~2/~1 (P)revious 2 exams s; 3 Is; 3 Is; 2 (randomly selected)

(R)CA ~1/~1/~1 (S)imulator

a. Shutdown Bank Withdrawal with Continuous Rod Motion The examinee will assume the shift in MODE 3 with prerequisites complete to withdraw Shutdown Bank "A" using O-S0-85-1 CONTROL ROD DRIVE SYSTEM to initialize the Rod Control Startup Step Counter Reset, close the Reactor Trip breakers and momentarily withdraw Shutdown Bank A.

When Shutdown Bank "A" is withdrawn a continuous rod motion condition occurs. The examinee will trip the Reactor using the Immediate Operator Actions of AOP-C.01, ROD CONTROL SYSTEM MALFUNCTIONS.

b. Depressurize the RCS to Enable Safety Injection Pump Flow During a Loss of Charging The examinee will assume the shift with the Reactor tripped and ES-0.1, Reactor Trip Response complete following a loss of all Charging pumps. The examinee will use AOP-M.09, LOSS OF CHARGING starting at step 27 to block Safety Injection and commence RCS Pressure reduction using a PORV to enable Safety Injection flow as demonstrated by Pressurizer level increasing.
c. Not examined
d. Start a Reactor Coolant Pump The examinee will assume the shift in MODE 3 with prerequisites complete to start a Reactor Coolant pump. The examinee will use 1-S0-68-2 REACTOR COOLANT PUMPS to start the Reactor Coolant Pump (RCP). Shortly after starting the pump, a high motor winding condition develops. The examinee will use Alarm Response 1-AR-M5-B and will transition to AOP-R.04 REACTOR COOLANT PUMP MALFUNCTIONS to manually trip the RCP.
e. Initiate AFW Flow to Steam Generators While in ECA-2.1 The examinee will assume the shift during a common fault that results in all Steam Generators depressurizing. The examinee will use ECA-2.1, UNCONTROLLED DEPRESSURIZATION OF ALL STEAM GENERATORS to close the TD AFW pump steam supply valve and to reduce AFW flow to each Steam generator to 50 .::!::. 10gpm.
f. Respond To High Containment Pressure Condition By Placing RHR Spray In Service.

The examinee will assume the shift during a LOCA and a High Containment Pressure condition. The examinee will use FR-Z.1, High Containment Pressure, starting at step 13 to establish RHR Spray flow to reduce Containment pressure. During the alignment, the B Train RHR Spray flow isolation valve FCV-72-41 fails to operate. The examinee uses the alternate path to re-establish B Train RHR to the injection mode and establishes A Train RHR to the Containment Spray mode.

g. Respond to a Loss of Shutdown Board The examinee will assume the shift in MODE 1 when a loss of power to the 1B 6.9kV Shutdown Board occurs. The examinee will use AOP-P.05 LOSS OF UNIT 1 SHUTDOWN BOARDS to evaluate RCP seal cooling and start 1A Charging pump. The examinee will evaluate the 1B Diesel generator and determine the Diesel requires an EMERGENCY STOP due to a failure of the cooling water valves to open. Finally, the examinee will stop all AFW flow to the SG from the TD AFW pump.
h. Respond to ERCW Pump Trip and Isolate the ERCW Leak The examinee will assume the shift in MODE 1 when a trip of the Q-A Pump occurs, the examinee will manually start the J-A ERCW pump using AOP-M.01, LOSS OF ESSENTIAL RAW COOLING WATER. After the pump is started, the examinee will evaluate ERCW flow and determine a rupture has occurred in "A" Train ERCW. The examinee will subsequently manually place all "A" Train ERCW pumps in PTL and open all alternate ERCW supply to DG valves.
i. Close MSIV's Locally The examinee will assume the shift with the Reactor tripped following a fault on the Unit 1 #1 Steam Generator with a failure of the MSIV's to close. The examinee will use EA-1-1, CLOSING MSIVs LOCALLY to simulate placing the Auxiliary Control Room transfer switches to AUX and subsequently simulate removing the control power fuses from 125V Vital Battery Boards I and II for the failed open MSIV.
j. Take Local Control of the Turbine Driven AFW Level Control valve LCV-3-175 The examinee will assume the shift following a Control Room evacuation with the Control Room staff unable to control Auxiliary Feed to the Unit 2 #4 Steam Generator. The examinee will use AOP-C.04 SHUTDOWN FROM AUXILIARY CONTROL ROOM, APPENDIX W.2 CONTROL OF UNIT 2 TURBINE DRIVEN AFW FLOW FROM OUTSIDE MCR to simulate aligning the Station Blackout air bottles to the failed open LCV-3-175, Unit 2 #4 Steam Generator level control valve and control Auxiliary Feed flow to the Unit 2 #4 Steam Generator. This JPM reflects a new plant modification.
k. Radiation Monitor O-RE-90-122 Flushing After Hi Radiation Signal Isolation.

The examinee will assume the shift following a high radiation isolation during Cask Decon Collector Tank release. The examinee will use O-S0-77-1 WASTE DISPOSAL SYSTEM (LIQUID), section 8.2 Radiation Monitor O-RE-90-122 Flushing After Hi Radiation Signal Isolation of Release to simulate flushing the Liquid Effluent Radiation Monitor in an attempt to clear the high radiation signal. The examinee will then simulate isolating the Cask Decon Collector Tank and flushing the Cask Decon Collector Tank discharge to the Floor Drain Collector tank. The examinee will then simulate re-aligning the Cask Decon Collector Tank for liquid release.

ES-301 Control Room/In-Plant Systems Outline Form ES-301-2 Facility: Sequoyah Nuclear Station 1 & 2 Date of Examination: 3/2/2015 Exam Level: RO D SRO-I D SRO-U .../ Operating Test No: 2015-301 Control Room Systems (8 for RO); (7 for SRO-I); (2 or 3 for SRO-U, including 1 ESF)

System I JPM Title Type Code* Safety Function

a. Continuous Rod Withdrawal While Withdrawing Shutdown Bank N, A, S 1 Rods During a Startup. APE 001 AA 1.05 (4.3/4.2)
b. Not Examined N/A N/A
c. Isolate the Ruptured Steam Generator with MSIV Failure to D,A,L,S 3 Close. EPE 038 EA 1.32 (4.6/4. 7)
d. Not Examined N/A N/A
e. Not Examined N/A N/A
f. Respond to High Containment Pressure, Place RHR Spray in D, A, EN, L, 5

Service. EPE E14 EA1 .1 (3.7/3.7) s

g. Not Examined N/A N/A
h. Not Examined N/A N/A In-Plant Systems (3 for RO); (3 for SRO-I); (3 or 2 for SRO-U)
i. Close MSIV's Locally. APE 040 AA1 .04 (4.3/4.3) N, E, L 4S
j. Take Local Control of the Turbine Driven AFW Level Control N, R, E, L 8 valve LCV-3-175. APE 068 AA1.03 (4.1/4.3)
k. Not Examined N/A N/A

@ All RO and SRO-I control room (and in-plant) systems must be different and serve different safety functions; all 5 SRO-U systems must serve different safety functions; in-plant systems and functions may overlap those tested in the control room.

  • Type Codes Criteria for RO I SRO-I I SRO-U (A)lternate path 4-6 I 4-6 I 2-3 (C)ontrol room (D)irect from bank S9/S8/S4 (E)mergency or abnormal in-plant <::1/<::1/<::1 (EN)gineered safety feature - I - I <::1 (control room system)

(L)ow-Power I Shutdown <::1/~1/<::1 (N)ew or (M)odified from bank including 1(A) <::2/<::2/<::1 (P)revious 2 exams  ::;; 3 I ::;; 3 I ::;; 2 (randomly selected)

(R)CA <::1/<::1/<::1 (S)imulator

a. Shutdown Bank Withdrawal with Continuous Rod Motion The examinee will assume the shift in MODE 3 with prerequisites complete to withdraw Shutdown Bank "A" using O-S0-85-1 CONTROL ROD DRIVE SYSTEM to initialize the Rod Control Startup Step Counter Reset, close the Reactor Trip breakers and momentarily withdraw Shutdown Bank A.

When Shutdown Bank "A" is withdrawn a continuous rod motion condition occurs. The examinee will trip the Reactor using the Immediate Operator Actions of AOP-C.01, ROD CONTROL SYSTEM MALFUNCTIONS.

b. Not Examined
c. Isolate the Steam Generator Tube Rupture (With MSIV Failure to Close)

The examinee will assume the shift with the Reactor tripped following a Steam Generator Tube Rupture. The examinee will use E-3, STEAM GENERATOR TUBE RUPTURE to attempt to isolate a ruptured steam generator using the ruptured steam generator MSIV. The ruptured steam generator MSIV will fail to close requiring the examinee will to isolate the steam paths of the ruptured Steam Generator by closing all intact Steam Generator MSIV's and the alternate flowpath isolation valves.

d. Not Examined
e. Not Examined
f. Respond To High Containment Pressure Condition By Placing RHR Spray In Service.

The examinee will assume the shift during a LOCA and a High Containment Pressure condition. The examinee will use FR-Z.1, High Containment Pressure, starting at step 13 to establish RHR Spray flow to reduce Containment pressure. During the alignment, the B Train RHR Spray flow isolation valve FCV-72-41 fails to operate. The examinee uses the alternate path to re-establish B Train RHR to the injection mode and establishes A Train RHR to the Containment Spray mode.

g. Not Examined
h. Not Examined
i. Close MSIV's Locally The examinee will assume the shift with the Reactor tripped following a fault on the Unit 1 #1 Steam Generator with a failure of the MSIV's to close. The examinee will use EA-1-1, CLOSING MSIVs LOCALLY to simulate placing the Auxiliary Control Room transfer switches to AUX and subsequently simulate removing the control power fuses from 125V Vital Battery Boards I and II for the failed open MSIV.
j. Take Local Control of the Turbine Driven AFW Level Control valve LCV-3-175 The examinee will assume the shift following a Control Room evacuation with the Control Room staff unable to control Auxiliary Feed to the Unit 2 #4 Steam Generator. The examinee will use AOP-C.04 SHUTDOWN FROM AUXILIARY CONTROL ROOM, APPENDIX W.2 CONTROL OF UNIT 2 TURBINE DRIVEN AFW FLOW FROM OUTSIDE MCR to simulate aligning the Station Blackout air bottles to the failed open LCV-3-175, Unit 2 #4 Steam Generator level control valve and control Auxiliary Feed flow to the Unit 2 #4 Steam Generator. This JPM reflects a new plant modification.
k. Not Examined

Appendix D Scenario Outline Form ES-D-1 Facility: Sequoyah Scenario No.: 1 Op Test No.: 2015-301 Examiners: Candidates: SRO ATC BOP Initial Conditions: 100% MOL, EOOS risk green, FT-1-3A is in MAINT BYPASS, RTS 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br />.

Turnover: Maintain 100% power, swap Condenser Vacuum pumps.

Critical Tasks CT-22 Manually close any open PZR PORV before completing step 6 of ECA-0.0.

CT-24 Manually energize at least one 6.9 kv Shutdown Board before placing safeguards equipment hand switches in the pull to lock position during ECA-0.0 (Step 10).

Event Event Type Event Description No.

1 (N)BOP/SRO The BOP will swap Condenser Vacuum pumps using 1-S0-2-9.

A low pressure FW heater String Isolation occurs, the BOP reduces turbine 2 ANOV179 (C)BOP/SRO load to maintain 10 minute average power less than 3455 MWt. (Admin Limit).

The crew reduces turbine load to less than 86% to comply with plant power 3 (R) ATC limitations.

(C) ATC #2 RCP #1 Seal fails, the ATC manually adjusts seal injection flow to 4 CV17B maintain > 9 gpm for RCP #4.

(C) SRO Power range Instrument N-44 fails high, the ATC takes immediate action to (I) ATC 5 NI07D place rod control in manual. The SRO declares power range instrument N-(l,TS) SRO 44 INOPERABLE.

ED05 (C) BOP A grid disturbance occurs resulting in excessive VAR loading on the main 6 generator, the BOP adjusts the main generator voltage regulator to bring EDR74 (C) SRO reactive load within limits.

The Transmission Operator informs the crew that the offsite power is 7 (TS) SRO DISQUALIFIED. The SRO will declare two offsite circuits INOPERABLE.

The grid continues to degrade and loss of offsite power occurs resulting in a 8 ED01 (M) ALL Reactor Trip. The crew responds using E-0 Reactor Trip or Safety Injection.

EG02B The B EOG trips and the A EOG fails to AUTO-START. The SRO 9 (C) BOP transitions to ECA-0.0.The BOP manually starts the 1A EOG by resetting EG03A the lockout relay to restore power to the 1A 6.9 Shutdown Board.

A power operated relief valve (PORV) will fail open; the ATC will isolate the 10 RC07A (C) ATC PORV using Prudent Operator Actions.

  • (N)ormal, (R)eactivitv, (l)nstrument, (C)omoonent, (M)ajor 1

Appendix D Scenario Outline Form ES-D-1

SUMMARY

Event 1 - The BOP will swap Condenser Vacuum pumps using 1-S0-2-9.

Event 2 - When directed by the Lead Examiner the "A" Low Pressure Feedwater Heater string isolation occurs. The crew will respond using AOP-S.04, Condensate or Heater Drains Malfunction. The BOP Reduces turbine load to maintain 10 minute average power less than 3455 MWt. (Administrative Limit)

Event 3 - The will perform a plant power reduction to <86% power using AOP-C.03, Rapid Shutdown or Load Reduction for the LP heater string isolation.

Event 4 - When directed by the Lead Examiner the #4 Reactor Coolant pump (RCP) The ATC will manually adjust RCP seal water supply flow> 9 gpm for RCP #4 using AOP-R.04 REACTOR COOLANT PUMP MALFUNCTIONS.

Event 5 - When directed by the Lead Examiner a power Range instrument Channel IV (N-44) fails high. The ATC places rod control in manual due to inadvertent insertion using immediate operator actions and AOP-C.01, ROD CONTROL SYSTEM MALFUNCTIONS. The crew will transition to AOP-1.01 NUCLEAR INSTRUMENT MALFUNCTION. The ATC will defeat the failed channel and place rod control to automatic. The SRO addresses Tech Specs and determines the power range instrument to be INOPERABLE and enters LCO 3.3.1.1 Action 2 and 6.

Event 6 - When directed by the Lead Examiner a grid disturbance occurs resulting in excessive VAR loading on the main generator, the BOP adjusts the main generator voltage regulator to bring reactive load within limits using AOP-P.07 DEGRADED GRID CONDITIONS OR GENERATOR VOLTAGE REGULATOR MALFUNCTION, APPENDIX D MVAR LIMITS FOR UNIT 1 GENERATOR STABILITY.

Event 7 - When directed by the Lead Examiner the Transmission Operator informs the crew that the offsite power is DISQUALIFIED. The SRO will enter AOP-P.07 DEGRADED GRID CONDITIONS OR GENERATOR VOLTAGE REGULATOR MALFUNCTION, APPENDIX A GENERIC ACTIONS FOR OFF-SITE POWER SOURCES INOPERABLE and enter LCO 3.8.1.1 Action D.

Event 8 - When directed by the Lead Examiner the grid continues to degrade and loss of offsite power occurs resulting in a Reactor Trip. The crew responds using E-0 Reactor Trip or Safety Injection.

Event 9 - The "B" EOG trips and the "A" EOG fails to AUTO-START. The SRO transitions to ECA-0.0 LOSS OF ALL AC POWER. The BOP manually starts the "A" EOG by resetting the lockout relay to restore power to the 1A 6.9 Shutdown Board. (Credit sought for a post trip component malfunction for the BOP. The verifiable action is an action that only the BOP will perform.)

Event 10 - When the Reactor trips a power operated relief valve (PORV) will fail open; the ATC will isolate the PORV using Prudent Operator Actions. (Credit sought for a post trip component malfunction for the ATC. The verifiable action is an action that only the ATC will perform.)

2

Appendix D Scenario Outline Form ES-D-1 Facility: Sequoyah Scenario No.: 2 Op Test No.: 2015-301 Examiners: Candidates: ATC SRO BOP Initial Conditions: 4% BOL, EOOS risk green FT-1-3A is in MAINT BYPASS, RTS 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br />.

Turnover: Continue the Reactor Startup, raise power to 13%

Critical Tasks CT-28 Manually align ECCS pump containment sump valves FCV-63-72 and FCV-63-73 to the Containment Sump prior to transition to ECA-1.1.

ES-1.3-PRA-2 Manually stop one of two Containment Spray Pumps while in ES-1.3 within 2 minutes of an RWST low level alarm of 27% during a SBLOCA or LBLOCA Event Event Type Event Description No.

1 (R)ATC Continue Power Increase from -4%.

1a (N)BOP/SRO Place feed regulating valves in AUTO using O-G0-4.Section 5.1 Step 4.

(C) BOP The A Hotwell pump develops a high current condition; the BOP will 2 CN01A (C)SRO manually start the C Hotwell pump.

The 1A CCS Pump trip coupled with a failure of the 1B-B CCS pump to auto CC09B (C) BOP start. The BOP will manually start the 1B CCS pump using AOP-M.03. The 3

SRO will address Tech Specs and determines the 1A CCS pump is CC10A (C TS)SRO INOPERABLE.

CV40 (C) ATC The Letdown Temperature Control valve fails to control in AUTO, the ATC 4

AUTO (C)SRO will control Letdown temperature in MANUAL.

(C) ATC A Small RCS Leak will develop. The ATC will raise charging flow using 5 TH01C AOP-R.05. The SRO will address Tech Specs and determine that RCS (C TS) SRO Leakage is in excess of TS requirements.

The RCS leak increases to a LBLOCA, the crew will initiate a Reactor trip 6 TH01C (M) ALL and Safety Injection and transitions to E-0.

The crew will transition from E-0 to E-1 and ultimately ES-1.3 to align RHR to the Containment Sump. During the alignment to the Containment sump, 7 RH02 (C) ATC a failure of sump AUTO swapover occurs. The ATC will manually align the suction of the ECCS pumps to the Containment Sump.

The A RHR pump trips and the B RHR pump will not start. The crew will 8 RH01 (C) ATC transition to ECA-1.1.

  • (N)ormal, (R)eactivitv, (l)nstrument, (C)omponent, (M)aior 1

Appendix D Scenario Outline Form ES-D-1

SUMMARY

Event 1 - The crew will assume the shift, place the feed regulating valves in AUTO using O-G0-4.Section 5.1 Step 4 and continue the power increase.

Event 2 - When directed by the Lead Examiner the "A" Hotwell pump current increases, the BOP will manually start the "C" Hotwell pump The crew may choose to use the guidance in either the ARP, AOP-S.04 or 1-S0-2/3-1 to start the "C" Hotwell pump.

Event 3 - When directed by the Lead Examiner the "A" CCS Pump trip coupled with a failure of the "B" CCS pump to auto start. The BOP will manually start the 1B-B CCS pump using prudent operator actions and AOP-M.03 LOSS OF COMPONENT COOLING WATER. The SRO will address Tech Specs and determines the "A" CCS pump is INOPERABLE and enters LCO 3. 7.3.

Event 4 - The Letdown Temperature Control valve fails to control in AUTO, the ATC will control Letdown temperature in MANUAL.

Event 5 - When directed by the Lead Examiner a small unisolable RCS leak will develop. The ATC will raise charging flow to control pressurizer level. The SRO will address Tech Specs and determine that RCS Leakage is in excess of LCO requirements. The SRO will enter LCO 3.4.6.2.a orb action a.

Event 6 - The RCS leak increases to a LBLOCA, the crew will initiate a Reactor trip and Safety Injection and transitions to E-0, REACTOR TRIP OR SAFETY INJECTION to trip the reactor and initiate Safety Injection and transition to E-1 LOSS OF REACTOR OR SECONDARY COOLANT. During the transition to E-1, status tree monitoring will occur. The crew will identify red path condition and implement FR-Z.1 for High Containment Pressure and potentially FR-P.1 for Pressurized Thermal Shock RED Path.

Event 7 - When the RWST depletes the crew will align ECCS to the Containment sump using ES-1.3 TRANSFER TO RHR CONTAINMENT SUMP. During the alignment to the Containment sump, a failure of sump AUTO swapover occurs. The ATC will manually align the suction of the ECCS pumps to the Containment Sump. (Credit sought for a post trip component malfunction for the ATC. The verifiable action is a Prudent Operator Action that only the ATC will perform.)

Event 8 - The "A" RHR pump trips and the "B" RHR pump will not start. The crew will transition to ECA-1.1 LOSS OF ECCS SUMP RECIRCULATION.

2

Appendix D Scenario Outline Form ES-D-1 Facility: Sequoyah Scenario No.: 3 V1- Op Test No.: 2015-301 Examiners: Candidates: ATC SRO BOP Initial Conditions: 100% BOL, EOOS risk green, FT-1-3A is in MAINT BYPASS, RTS 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br />.

Turnover: Maintain 100% power Critical Tasks CT-18 Manually isolates feed water flow into and steam flow from the ruptured SIG before a transition to ECA-3.1 occurs.

CT-19 Manually establish/maintain an RCS temperature so that transition from E-3 does not

.occur because the RCS temperature is in either of the following conditions:

Too high to maintain < 60 degrees CET sub cooling .

.OR either of the following RED path on PTS Status Tree

.OR ORANGE path on Sub criticality Status Tree CT-20 Manually depressurize RCS to meet and maintain SI termination criteria while in E-3 prior to exceeding 100% SIG dome level on the simulator during a SGTR of 430 gpm or less (TH05 9.15%).

Event Event Type Event Description No.

TC10RLY2 The# 3 Heater Drain Tank level control valve fails open resulting in a 1 (C)BOP/SRO run back signal with a failure of the Main Turbine. The BOP will initiate a HD06A Main Turbine runback using the valve position limiter.

Following the manual runback, a malfunction causes the control rods (C) ATC continue to insert beyond the normal position, the ATC places control rods 2 RD17 (C,TS) SRO in manual using Immediate Operator Actions and will initiate emergency boration. The SRO enters the Rod Insertion Limit LCO.

A small Steam Generator Tube Leak develops on the #1 SG; the ATC will (C) ATC 3 TH05A raise Charging flow to maintain Pressurizer Level on program. The SRO will (TS,C)SRO enter the RCS Leakage LCO.

The SGBD isolation valve fails to close; the BOP will manually isolate 4 RC13 (C)BOP/SRO SGBD flow by closing the SGBD flow control valve.

(R) ATC 5 The crew will reduce power in response to the SGTL.

(N)BOP/SRO The tube leak will degrade to a SGTR; the crew will trip the reactor, initiate 6 TH05A (M) ALL Safety Injection and enter E-0. The crew will transition to E-3 to, isolate and cool down the ruptured SG and depressurize the RCS.

The #1 SG Atmospheric Dump fails open when the reactor trips. The BOP 7 MS12A (C) BOP manually closes the Atmospheric Dump using prudent operator actions.

  • (N)ormal, (R)eactivitv, (l)nstrument, (C)omponent, (M)aior 1

Appendix D Scenario Outline Form ES-D-1

SUMMARY

Event 1 - When directed by the Lead Examiner, the# 3 Heater Drain Tank level control valve fails open resulting in a runback signal. The BOP initiates a Main Turbine runback using the valve position limiter using AOP-S.04 CONDENSATE OR HEATER DRAINS MALFUNCTION.

Event 2 - Following the manual runback, a malfunction causes the control rods continue to insert beyond the normal position; the ATC places control rods in manual using Immediate Operator Actions and AOP-C.01 ROD CONTROL SYSTEM MALFUNCTIONS to stop the excessive insertion. The SRO enters the Control Bank Rod Insertion Limit LCO. The ATC will initiate emergency boration to aid in compliance with the LCO.

Event 3 - When directed by the lead examiner, SG Tube Leak develops on the #1 SG. The ATC will manually increase charging flow and subsequently isolate letdown using AOP-R.01 STEAM GENERATOR TUBE LEAK. The SRO will enter LCO 3.4.6.2.c Action a.

Event 4 - Blow down Isolation valve 15-44 fails to auto-isolate, the BOP will isolate Blow down flow by closing the SG Blow down Flow Control valve.

Event 5 -The crew will reduce power in response to the SGTL using AOP-C.03 RAPID SHUTDOWN OR LOAD REDUCTION.

Event 6 - The tube leak will degrade to a SGTR; the crew will trip the reactor, initiate Safety Injection and enter E-0, REACTOR TRIP OR SAFETY INJECTION. The crew will isolate and cool down the ruptured SG and depressurize the RCS using E-3, STEAM GENERATOR TUBE RUPTURE.

Event 7 - The #1 SG Atmospheric Dump fails open when the reactor trips. The BOP manually closes the Atmospheric Dump using prudent operator actions. (Credit sought for a post trip component malfunction for the BOP. The verifiable action is a Prudent Operator Action that only the BOP will perform.)

2

Appendix D Scenario Outline Form ES-D-1 Facility: Sequoyah Scenario No.: 3 V2 Op Test No.: 2015-301 Examiners: Candidates: ATC SRO BOP Initial Conditions: 100% BOL, EOOS risk green, FT-1-3A is in MAINT BYPASS, RTS 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br />.

Turnover: Maintain 100% power Critical Tasks CT-18 Manually isolates feed water flow into and steam flow from the ruptured SIG before a transition to ECA-3.1 occurs.

CT-19 Manually establish/maintain an RCS temperature so that transition from E-3 does not occur because the RCS temperature is in either of the following conditions:

Too high to maintain < 60 degrees CET sub cooling .

.OR either of the following RED path on PTS Status Tree

.OR ORANGE path on Sub criticality Status Tree CT-20 Manually depressurize RCS to meet and maintain SI termination criteria while in E-3 prior to exceeding 100% SIG dome level on the simulator during a SGTR of 430 gpm or less (TH05 9.15%).

Event Event Type Event Description No.

A small Steam Generator Tube Leak develops on the #1 SG; the ATC will (C) ATC 1 TH05A raise Charging flow to maintain Pressurizer Level on program. The SRO will (TS,C)SRO enter the RCS Leakage LCO.

The SGBD isolation valve fails to close; the BOP will manually isolate 2 RC13 (C)BOP/SRO SGBD flow by closing the SGBD flow control valve.

(R) ATC 3 The crew will reduce power in response to the SGTL.

(N)BOP/SRO TC10RLY2 The# 3 Heater Drain Tank level control valve fails open resulting in a 4 (C)BOP/SRO run back signal with a failure of the Main Turbine. The BOP will initiate a HD06A Main Turbine runback using the valve position limiter.

Following the manual runback, a malfunction causes the control rods (C) ATC continue to insert beyond the normal position, the ATC places control rods 5 RD17 (C,TS) SRO in manual using Immediate Operator Actions and will initiate emergency boration. The SRO enters the Rod Insertion Limit LCO.

The tube leak will degrade to a SGTR; the crew will trip the reactor, initiate 6 TH05A (M) ALL Safety Injection and enter E-0. The crew will transition to E-3 to, isolate and cool down the ruptured SG and depressurize the RCS.

The #1 SG Atmospheric Dump fails open when the reactor trips. The BOP 7 MS12A (C) BOP manually closes the Atmospheric Dump using prudent operator actions.

  • (N)ormal, (R)eactivitv, (l)nstrument, (C)omponent, (M)ajor 1

Appendix D Scenario Outline Form ES-D-1

SUMMARY

Event 1 - When directed by the lead examiner, SG Tube Leak develops on the #1 SG. The ATC will manually increase charging flow and subsequently isolate letdown using AOP-R.01 STEAM GENERATOR TUBE LEAK. The SRO will enter LCO 3.4.6.2.c Action a.

Event 2 - Blow down Isolation valve 15-44 fails to auto-isolate, the BOP will isolate Blow down flow by closing the SG Blow down Flow Control valve.

Event 3 -The crew will reduce power in response to the SGTL using AOP-C.03 RAPID SHUTDOWN OR LOAD REDUCTION.

Event 4 -When directed by the Lead Examiner, the# 3 Heater Drain Tank level control valve fails open resulting in a runback signal. The BOP initiates a Main Turbine runback using the valve position limiter using AOP-S.04 CONDENSATE OR HEATER DRAINS MALFUNCTION.

Event 5 - Following the manual runback, a malfunction causes the control rods continue to insert beyond the normal position; the ATC places control rods in manual using Immediate Operator Actions and AOP-C.01 ROD CONTROL SYSTEM MALFUNCTIONS to stop the excessive insertion. The SRO enters the Control Bank Rod Insertion Limit LCO. The ATC will initiate emergency boration to aid in compliance with the LCO.

Event 6 - The tube leak will degrade to a SGTR; the crew will trip the reactor, initiate Safety Injection and enter E-0, REACTOR TRIP OR SAFETY INJECTION. The crew will isolate and cool down the ruptured SG and depressurize the RCS using E-3, STEAM GENERATOR TUBE RUPTURE.

Event 7 - The #1 SG Atmospheric Dump fails open when the reactor trips. The BOP manually closes the Atmospheric Dump using prudent operator actions. (Credit sought for a post trip component malfunction for the BOP. The verifiable action is a Prudent Operator Action that only the BOP will perform.)

2

Appendix D Scenario Outline Form ES-D-1 Facility: Sequoyah Scenario No.: 4 Op Test No.: 2015-301 Examiners: Candidates: ATC SRO BOP Initial Conditions: 95% BOL, EOOS risk green, FT-1-3A is in MAINT BYPASS, RTS 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br />.

Turnover: Raise power to 100%

Critical Tasks CT-7 Manually start at least one required high-head (Safety Injection) ECCS pump prior to transition out of E-0 during a SBLOCA when RCS pressure is less than 1500 psig.

CT-16 Manually trip all Reactor Coolant Pumps given the following conditions:

One Safety Injection pump is not functioning.

AND One of the following malfunctions is inserted:

. A 2" RCS Cold Leg break (TH02 0.529%) for greater than 33 minutes .

OR

. A 3" RCS Cold Leg break (TH02 1.19%) for greater than 10 minutes but less than 13.5 minutes.

AND The RCP's automatically trip after the above times has elapsed.

Event Event Type Event Description No.

(R) ATC Raise Reactor power to 100%

1 (N)BOP/SRO ED120B1B 120 V vital power breaker to lower Containment radiation monitor 1-RM 2 TS-SRO 007 106 trips. The SRO declares the radiation Monitor INOPERABLE.

The 3A Heater Drain Pump trips which results in a Main Turbine runback; 3 RD09 (C) ATC/SRO the ATC manually inserts control rods due to a failure of the Rod Control system.

A Pressurizer Level instrument failure occurs resulting in a loss of Letdown.

(I) ATC/SRO 4 RX06A The ATC will de-select the affected instrument and restore Letdown. The (TS) SRO SRO declares the instrument INOPERABLE.

(C) BOP A failure will occur on #1 SG Steam Flow Instrumentation. The BOP will 5 RX09VE (C) SRO place the Feed Reg Valve to #1 SG in MANUAL and control SG level.

A leak will develop in the Condenser Cooling water system in the Turbine (C) BOP 6 ANOV699 Building. The crew will trip the Reactor. The BOP will stop all Unit 1 CCW (C)SRO pumps and isolate all Unit 1 Main Condenser Waterboxes to stop the leak.

A SBLOCA will develop. The crew will respond using E-0 to initiate Safety 7 TH02A (M) ALL Injection and ultimately transition to E-1.

During the safety Injection, the A SI pump will trip and the B SI pump will fail 8 Sl13B (C) ATC to AUTO start, the ATC will manually start the B SI pump.

  • (N)ormal, (R)eactivitv, (l)nstrument, (C)omponent (M)aior 1

Appendix D Scenario Outline Form ES-D-1

SUMMARY

Event 1 - The crew raises plant power using G0-5 section 5.1 from 95% to 100% power Event 2 - When directed by the lead examiner, 120 V vital power breaker to lower Containment radiation monitor 1-RM-90-106 trips. The SRO will enter LCO 3.3.3.1 action 27 and LCO 3.4.6.1 action b.

Event 3 - When directed by the lead examiner, the 3A Heater Drain Pump trips resulting in a Main Turbine run back. The ATC manually inserts control rods due to a failure of the Rod Control system using prudent operator actions.

Event 4 - When directed by the lead examiner, the pressurizer level channel LT 68-339 will fail low resulting in letdown isolation and de-energizing Pressurizer heaters. The ATC will remove the channel from service and restore Letdown using AOP-1.04, PRESSURIZER INSTRUMENT AND CONTROL MALFUNCTIONS. The SRO will enter LCO 3.3.1.1 Action 6 and LCO 3.3.3.7 Action 2.

Event 5 - When directed by the lead examiner a failure will occur on #1 SG Steam Flow Instrumentation. The BOP will place the Feed Regulating Valve to #1 SG in MANUAL and control SG level using AOP-S.01 MAIN FEEDWATER MALFUNCTIONS.

Event 6 - A leak will develop in the Condenser Cooling water system in the Turbine Building.

The crew will trip the Reactor. The BOP will stop all Unit 1 CCW pumps and isolate all Unit 1 Main Condenser Water boxes to stop the leak using AOP-M.08 INTERNAL FLOODING.

Event 7 - When directed by the lead examiner, a SBLOCA will develop. The crew will respond using E-0, REACTOR TRIP OR SAFETY INJECTION to initiate Safety Injection and ultimately transition to E-1 LOSS OF REACTOR OR SECONDARY COOLANT.

Event 8 - During the safety Injection, the A SI pump will trip and the B SI pump will fail to AUTO start, the ATC will manually start the B SI pump. (Credit sought for a post trip component malfunction for the ATC. The verifiable action is a Prudent Operator Action that only the ATC will perform.)

2

Appendix D Scenario Outline Form ES-D-1 Facility: Sequoyah Scenario No.: 5 Op Test No.: 2015-301 Examiners: Candidates: ATC SRO BOP Initial Conditions: 48% BOL, EOOS risk green, FT-1-3A is in MAINT BYPASS, RTS 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br />.

Turnover: Start the B Main Feed Pump, then raise% power to 100%.

Critical Tasks CT-13 Manually trip the main turbine before an orange path challenge develops to either the sub criticality (S) or the Pressurized Thermal Shock (P) CSF or before transition to ECA-2.1, whichever happens first.

CT-45 Manually establish at least 440 gpm feed water flow rate to the SGs while in FR-H.1 before any three SG wide range levels are less than 23% [43% ADV].

Event Event Type Event Description No.

Injection Water Pump A Trips with a failure of the B Injection Water pump to CN12B (C) BOP 1 AUTO Start. The BOP manually starts the B Injection Water pump using the (C)SRO ARP.

Impulse pressure transmitter PT-1-73 fails high resulting in rod withdrawal.

RX11B (I) ATC 2 The ATC places control rods in manual using Immediate Operator Actions.

(TS,I) SRO The SRO declares PT-1-73 INOPERABLE.

Steam Gen Level Transmitter, LT 3-94 Fails low, the crew will enter AOP 3 RX16C TS-SRO 1.06. The SRO will address Tech Specs and determines the instrument is INOPERABLE.

An Exciter Insulation Resistance Low alarm is received. The crew will AN_OV_59 (R) ATC 4 reduce power in response to the Main Generator alarm or from plant (N)BOP/SRO management direction.

Upon the initiation of the reduction in power, normal boration controller FIC-CV38 (C) ATC 5 62-139 fails closed, the ATC initiates Emergency Boration to facilitate the (C)SRO reduction of power.

During the of the reduction in power, Feed Regulating valve FCV-3-48 fails FW16B (C) BOP 6 to control in AUTO, the BOP places FIC-3-48 in MANUAL to control SG (C)SRO level on program.

7 FWR13A (M) ALL The A Main Feed water Pump trips, the crew manually trips the Reactor.

TC11ALL The Main Turbine fails to AUTO trip; the BOP manually trips the Main 8 TC12ALL (C) BOP Turbine during Immediate Operator Actions.

FW07 All Auxiliary Feed water pumps fail. The SRO transitions to FR-H.1 and 9 (C) BOP establishes Feed water using a Main Feed water pump.

  • (N)ormal, (R)eactivity (l)nstrument, (C)omponent, (M)aior 1

Appendix D Scenario Outline Form ES-D-1

SUMMARY

Event 1 - When directed by the Lead Examiner, Injection Water Pump A Trips with a failure of the B Injection Water pump to AUTO Start. The BOP manually starts the B Injection Water pump using the ARP.

Event 2 - When directed by the Lead Examiner, turbine first stage pressure transmitter, PT 73, will fail low. The ATC will place control rods to MANUAL to stop the continuous rod withdrawal using Immediate Operator Actions of AOP-C.01, Rod Control System Malfunctions.

The crew will transition to AOP-1.08, Turbine Impulse Pressure Instrument Malfunction to address the RCS temperature control, feed water control and steam dump realignment. The SRO will enter LCO 3.3.1.1 Action 8.b.

Event 3 - Steam Gen Level Transmitter, LT 3-94 fails low, the SRO will enter AOP 1.06 AOP-1.06 STEAM GENERATOR INSTRUMENT MALFUNCTION. The SRO will enter Enters LCO 3.3.1.1Action9, LCO 3.3.2.1Action17, Action 22c, and Action 36 and enters LCO 3.3.3.7.

Action 1.

Event 4 - An Exciter Insulation Resistance Low alarm is received. The crew will reduce power in response to the Main Generator alarm or from plant management direction using AOP-C.03 RAPID SHUTDOWN OR LOAD REDUCTION.

Event 5 - Upon the initiation of the reduction in power, normal boration controller FIC-62-139 fails closed; the ATC initiates Emergency Boration using AOP-C.03 RAPID SHUTDOWN OR LOAD REDUCTION, APPENDIX I EMERGENCY BORATION to facilitate the reduction of power.

Event 6 - During the of the reduction in power, Feed Regulating valve FCV-3-48 fails to control in AUTO, the BOP places FIC-3-48 in MANUAL to control SG level on program using the Immediate Operator Actions of AOP-S.01 MAIN FEEDWATER MALFUNCTIONS.

Event 7 - The "A" Main Feed water pump trips, the crew manually trips, the Reactor and enters E-0, REACTOR TRIP OR SAFETY INJECTION.

Event 8 - The Main Turbine fails to AUTO trip; the BOP manually trips the Main Turbine during Immediate Operator Actions of E-0, REACTOR TRIP OR SAFETY INJECTION. (Credit sought for a post trip component malfunction for the BOP. The verifiable action is an Immediate Operator Action that only the BOP will perform.)

Event 9 - All Auxiliary Feed water pumps fail. The SRO transitions to FR-H.1 LOSS OF SECONDARY HEAT SINK and establishes Feed water using a Main Feed water pump. (Credit sought for a post trip component malfunction for the BOP. The verifiable action is an action that only the BOP will perform.)

2

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